The present invention relates to a measurement method of an axial void fraction distribution in a core of a boiling water reactor (BWR), and also to an evaluating method of a neutron multiplication factor of a fuel assembly to be contained in a container apparatus.
In a BWR, cooling water is heated and boiled in a reactor core as it flows from the bottom of the core to the top of the core. So, a void fraction that is a ratio of the bubbles to space where the cooling water flows in a channel box, increases as the cooling water flows from the upstream (the bottom of the core) to the downstream (the top of the core).
The void fraction significantly affects core characteristics such as reactivity, power distribution and cooling characteristics. Therefore, it is important to quantitatively evaluate the void fraction distribution.
However, no report has been made on an actual measurement of the void fraction distribution in the core of a commercial BWR because of lack of feasible measurement technique. Conventionally, a mock experiment is conducted out of the commercial reactor to build a theoretical model, and based upon the model, the void fraction distribution of the actual BWR is computationally evaluated.
It has been recognized that the measurement of the void fraction distribution is necessary for a long time. It was reported that a change of nuclear characteristics due to a change of void fraction was indirectly measured by simulating voids in a critical facility that rarely emit thermal power.
In this measurement, a wire containing manganese (Mn), called Manganin (Trade Mark) wire was installed in an experimental reactor core, and a thermal neutron absorber, cadmium (Cd), was wound around the Manganin wire. A cadmium ratio (Cd-ratio) in Mn55 (n, γ) reactions was measured.
In this measurement, a wire containing manganese (Mn) (Manganin wire) was partly wrapped with cadmium (Cd), which is a thermal neutron absorber, and introduced into the reactor core of the experiment system. Then, the cadmium ratio in the reaction ratio of the reaction of Mn55 (n, γ) was measured.
Various diameters of aluminum tubes were introduced in a test core and Cd-ratios were measured. These measurements show that Cd-ratio has a good correlation with a void fraction. So the report says that a void fraction can be measured by measuring Cd-ratio. The basic principle of the proposed technique is excellent, because it is based on the fact that a void fraction has a good correlation with the ratio of the non-thermal neutron flux and the thermal neutron flux.
One of the inventors of the present invention proposed more practical techniques that use the basic principle described above but does not require the use of Cd in Japanese Patent Application Disclosure No. Sho 55-121195 and Japanese Patent Application Disclosure No. Sho 55-125489. According to these documents, a strong thermal neutron absorber such as Gd2O3 that has a high melting point or a weak neutron absorber such as stainless steel is locally arranged in fuel assemblies, in-core instrumentation tube, a fixed position neutron detector or a movable neutron detector to cause a local distortion in a thermal neutron flux. Then, a thermal neutron flux and a non-thermal neutron flux are separated, and a void fraction is determined from a ratio of the two neutron fluxes.
The above-described technique is hard to apply to an operating nuclear reactor, because it employs Cd having a low melting point that might harm structure materials of the operating reactor.
The inventors of the present invention have been studying new techniques of utilizing an in-core instrumentation tube. One of the techniques is a void fraction measurement method using a ratio of a non-thermal flux and a thermal flux, as well as a ratio of a thermal neutron flux and a fast neutron flux that has a larger dependency on a void fraction than non-thermal neutron flux.
To use this technique, it is necessary to discharge the in-core instrumentation tube from the reactor and measure the tube during an outage of reactor operation. At the time of measurement, a dedicated holder has to be prepared in order to accurately place the instrumentation tube and the detector because the shape of the tube has to be maintained finely and accurately for measurement. In addition, since the instrumentation tube is arranged in the water gap between the fuel assemblies, a measured void fraction is an average of at least four surrounding fuel assemblies, and the instrumentation tube and the detection sensitivity is significantly lower than inside of a fuel assembly.
As described above, while the axial void fraction distribution is very important for a BWR, no technique has been available for actual measurement of an axial void fraction distribution in an operating commercial nuclear reactor.
When discharging an irradiated BWR fuel assembly from the core and containing it in a fuel assembly containing apparatus such as a transport vessel (cask) or a fuel storage rack in water, a neutron multiplication factor (or reactivity) at the axial position of ⅔ to ¾ from the lower end of the axial fuel active part tends to be higher. Such a trend appears due to design requirements such as that the flat axial thermal power distribution is preferable during the operation and due to a delay of burning of uranium and generated plutonium in an upper part (downstream of the coolant flow) because of an effect of the axial void fraction distribution and of a high generation rate of plutonium as a result of a high conversion ratio.
Under these circumstances, for a purpose of an assured sub-criticality analysis, i.e. making sure of criticality safety, it is necessary to take an axial void fraction distribution into account to evaluate a multiplication factor. One way to achieve this purpose is to include a large design margin.
In view of the above-identified problems, it is therefore the first object of the present invention to provide a novel and highly feasible method to evaluate an axial void distribution. The second object of the present invention is to provide a method to experimentally evaluate the neutron multiplication factor of a fuel assembly so that a large design margin may not be required to avoid a critical accident while containing an irradiated (discharged) BWR fuel assembly in a fuel assembly containing apparatus.